• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.364-372
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• 2023

The space reactor is the primary energy supply for future space vehicles and space stations. The radiator is one of the essential parts of a space reactor. Therefore, the research on radiators can improve the heat dissipation power, reduce the quality of radiators, and make the space reactor smaller. Based on MOOSE multi-physics numerical calculation platform, a simulation program for the combination of heat pipe and fin at the end of heat pipe radiator is developed. It is verified that the calculation result of this program is accurate and the calculation speed is fast. Analyze the heat transfer characteristics of the combination with heat pipe and fin, and obtain its internal temperature field. Based on the calculation results, the influence of structural parameters on the heat dissipation power is analyzed. The results show that when the fin width is 0.25 m, fin thickness is 0.002 m, condensing section length is 0.5425 m and heat pipe radius is 0.014 m, the power-mass ratio is the highest. When the temperature is 700K-900K, the heat dissipation power increases 41.12% for every 100K increase in the operating temperature. Smaller fin width and thinner fin thickness can improve the power-mass ratio and reduce the radiator quality.

• Journal of the Korean Society of Safety
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• v.35 no.4
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• pp.84-91
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• 2020

The Korean nuclear industry developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code and this code adpots two-phase flows, two-fluid, three-field models which are comprised of gas, continuous liquid and droplet fields and has a capability to simulate three-dimensional model. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for accident management plan of nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification work for separate and integral effect experiments is required. In this reason, the goal of this work is to verify calculation capability of SPACE code for multiple failure accident. For this purpose, it was selected the experiment which was conducted to simulate a Multiple Steam Generator Tube Rupture(MSGTR) accident with Passive Auxiliary Feedwater System(PAFS) operation by Korea Atomic Energy Research Institute (KAERI) and focused that the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The MSGR accident has a unique feature of the penetration of the barrier between the Reactor Coolant System (RCS) and the secondary system resulting from multiple failure of steam generator U-tubes. The PAFS is one of the advanced safety features with passive cooling system to replace a conventional active auxiliary feedwater system. This system is passively capable of condensing steam generated in steam generator and feeding the condensed water to the steam generator by gravity. As the results of overall system transient response using SPACE code showed similar trends with the experimental results such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it could be concluded that the SPACE code has sufficient capability to simulate a MSGTR accident.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2534-2546
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• 2023

Gas-cooled space reactor, which adopts He-Xe gas mixture as working fluid, is a better choice for megawatt power generation. In this paper, thermal-hydraulic characteristics of He-Xe gas mixture in 2×2 rod bundle wrapped with helical wires is numerically investigated. The velocity, pressure and temperature distribution of the coolant are obtained and analyzed. The results show that the existence of helical wires forms the vortexes and changes the velocity and temperature distribution. Hot spots are found at the contact corners between helical wires and fuel rods. The highest temperature of the hot spots reach 1600K, while the mainstream temperature is less than 400K. The helical wire structure increases the friction pressure drop by 20%-50%. The effect extent varies with the pitch and the number of helical wires. The helical wire structure leads to the reduction of Nusselt number. Comparing thermal-hydraulic performance ratios (THPR) of different structures, the THPR values are all less than 1. It means that gas-cooled space reactor adopting helical wires could not strengthen the core heat removal performance. This work provides the thermal-hydraulic design basis for He-Xe gas cooled space nuclear reactor.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.3039-3045
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• 2023

Fiber-reinforced polymer (FRP) composites are considered suitable candidates for structural materials of spacecrafts due to their excellent properties of high strength, light weight, and corrosion resistance. An online health monitoring method for FRP composites must be applied to space structures. However, the application of existing health monitoring methods to space structures is limited due to the harsh space environment. Here, carbon fiber-reinforced polymer (CFRP) composites embedded with fiber Bragg grating (FBG) sensors were prepared to explore the feasibility of strain monitoring using embedded FBG sensors in γ-radiation environment. The analysis of the influence of radiation on the strain monitoring demonstrated that the embedded FBG can be successfully applied to the health monitoring of FRP composites in radiation environment.

• The Korean Journal of Air & Space Law and Policy
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• v.32 no.2
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• pp.219-246
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• 2017

Nuclear issue is a good example showing globalization of the international regime. The history showed nuclear weapons may cause the extinction of human races when the first nuclear bombs fell down to Japan in August, 1945 and people became increasingly eager to achieve peace. Military buildup for national security is a matter of existence in the international society. However, disarmament or arms control to secure international peace and safety which is also the purpose of the Charter of the United Nations may be the most important task for us to realize peace of the mankind. Today, disarmament, together with amicable settlement of international conflicts and collective security system, is an important means to maintain and promote international peace and safety. It might be our permanent task to realize complete disarmament but, as the Preamble of the Treaty on the Non-Proliferation of Nuclear Weapons (NPT) expressed general and complete disarmament, the international society has clarified its effort for complete disarmament. Thus, taking a look into the international regime on the nuclear issue and progress related to the nuclear issue in North Korea, the study was intended to introduce the globalization of the nuclear issue, review the international effort for nuclear disarmament based on the concept of the 'common heritage of the mankind' and with respect to the United Nations Committee on the Peaceful Uses of Outer Space (COPUOS) and controls over nuclear weapons, and then evaluate the North Korean nuclear issue, which is in direct relation with South Korea and international laws, in terms of the space law and disarmament acts. The collective security system along with policies to prevent dissemination of nuclear weapons should also be emphasized and implemented to cope with the North Korean nuclear issue.

• Korean Journal of Mathematics
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• v.9 no.2
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• pp.83-98
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• 2001

We characterize ($r,w$)-nuclear operators acting from a Banach space whose dual has weak type $q$ into a Banach space of weak type $p$ by the asymptotic behaviour of their entropy numbers.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.134-140
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• 2017

A well-performed core power control to track load changes is crucial in pressurized water reactor (PWR) nuclear power stations. It is challenging to keep the core power stable at the desired value within acceptable error bands for the safety demands of the PWR due to the sensitivity of nuclear reactors. In this paper, a state-space model predictive control (MPC) method was applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, the MPC model, and quadratic programming (QP). The mathematical models of the reactor core were based on neutron dynamic models, thermal hydraulic models, and reactivity models. The MPC model was presented in state-space model form, and QP was introduced for optimization solution under system constraints. Simulations of the proposed state-space MPC control system in PWR were designed for control performance analysis, and the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4776-4797
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• 2022

In the one-dimensional thermal-hydraulic (TH) codes, a subcooled boiling model to predict the void fraction profiles in a vertical channel consists of wall heat flux partitioning, the vapor condensation rate, the bubbly-to-slug flow transition criterion, and drift-flux models. Model performance has been investigated in detail, and necessary refinements have been incorporated into the Safety and Performance Analysis Code (SPACE) developed by the Korean nuclear industry for the safety analysis of pressurized water reactors (PWRs). The necessary refinements to models related to pumping factor, net vapor generation (NVG), vapor condensation, and drift-flux velocity were investigated in this study. In particular, a new NVG empirical correlation was also developed using artificial neural network (ANN) techniques. Simulations of a series of subcooled flow boiling experiments at pressures ranging from 1 to 149.9 bar were performed with the refined SPACE code, and reasonable agreement with the experimental data for the void fraction in the vertical channel was obtained. From the root-mean-square (RMS) error analysis for the predicted void fraction in the subcooled boiling region, the results with the refined SPACE code produce the best predictions for the entire pressure range compared to those using the original SPACE and RELAP5 codes.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.888-895
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• 2022

Lately, ocean nuclear power plants have attracted attention as one of diverse uses of nuclear power plants. Because ocean nuclear power plants are movable or transportable, it is necessary to analyze the thermal hydraulics in a moving frame of reference, and computer codes have been developed to predict thermal hydraulics in large moving systems. The purpose of this study is to incorporate a three dimensional dynamic motion model into the SPACE code (Safety and Performance Analysis CodE) so that the code is able to analyze thermal hydraulics in an ocean nuclear power plant. A rotation system that describes three-dimensional rotations about an arbitrary axis was implemented, and modifications were made to the one-dimensional momentum equations to reflect the rectilinear and rotational acceleration effects. To demonstrate the code's ability to solve a problem utilizing a rotational frame of reference, code calculations were conducted on various conceptual problems in the two-dimensional and three-dimensional pipeline loops. In particular, the code results for the three-dimensional pipeline loop with a tilted rotation axis agreed well with the multi-dimensional CFD results.

• Nuclear Engineering and Technology
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• v.37 no.2
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• pp.127-138
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• 2005

Approaches to several recent issues in the operation of nuclear power plants using computational intelligence are discussed. These issues include 1) noise analysis techniques, 2) on-line monitoring and sensor validation, 3) regularization of ill-posed surveillance and diagnostic measurements, 4) transient identification, 5) artificial intelligence-based core monitoring and diagnostic system, 6) continuous efficiency improvement of nuclear power plants, and 7) autonomous anticipatory control and intelligent-agents. Several changes to the focus of Computational Intelligence in Nuclear Engineering have occurred in the past few years. With earlier activities focusing on the development of condition monitoring and diagnostic techniques for current nuclear power plants, recent activities have focused on the implementation of those methods and the development of methods for next generation plants and space reactors. These advanced techniques are expected to become increasingly important as current generation nuclear power plants have their licenses extended to 60 years and next generation reactors are being designed to operate for extended fuel cycles (up to 25 years), with less operator oversight, and especially for nuclear plants operating in severe environments such as space or ice-bound locations.